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Kunieda, Satoshi; Iwamoto, Osamu; Iwamoto, Nobuyuki; Minato, Futoshi; Okamoto, Tsutomu; Sato, Tatsuhiko; Nakashima, Hiroshi; Iwamoto, Yosuke; Iwamoto, Hiroki; Kitatani, Fumito; et al.
JAEA-Conf 2016-004, p.41 - 46, 2016/09
Neutron- and proton-induced cross-section data are required in a wide energy range beyond 20 MeV, for the design of accelerator applications. New evaluations are performed with recent knowledge in the optical and pre-equilibrium model calculations. We also evaluated cross-sections for p+
Li and p+
Be which have been highly requested from a medical field. The present high-energy nuclear data library, JENDL-4.0/HE, includes evaluated cross-sections for incident neutrons and protons up to 200 MeV (for about 130 nuclei). We overview substantial features of the library, i.e., (1) systematic evaluation with CCONE code, (2) challenges for evaluations of light nuclei and (3) inheritance of JENDL-4.0 and JENDL/HE-2007. In this talk, we also focus on the results of benchmark calculation for neutronics to show performance of the present library.
Yamano, Naoki*; Iwamoto, Osamu; Nakamura, Shoji; Kunieda, Satoshi; Van Rooijen, W.*; Koura, Hiroyuki
JAEA-Conf 2014-002, 209 Pages, 2015/02
JAEA-Conf-2014-002.pdf:64.24MB
The 2013 Symposium on Nuclear Data was held at Research Institute of Nuclear Engineering, University of Fukui, on 14th and 15th of November 2013. The Nuclear Data Division of the Atomic Energy Society of Japan and Research Institute of Nuclear Engineering, University of Fukui organize this symposium in cooperation with Nuclear Science and Engineering Center of Japan Atomic Energy Agency and the Chubu Branch of Atomic Energy Society of Japan. In the oral sessions, papers were presented on topics of progress in neutron cross-section measurement and analysis, application of nuclear data, recent topics on nuclear data measurement and theory, and progress in studies of high-energy nuclear reactions. In the poster session, papers were presented concerning experiments, evaluations, benchmark tests and applications. This report consists of total 35 papers including 14 oral presentations and 21 poster presentations.
Amano, Osamu
Nihon Genshiryoku Gakkai-Shi ATOMO
, 56(3), p.14 - 15, 2014/03
Along with the rapid increase of the number of plants of the 1970s and 1980s, safety culture is degraded, it is in the lowest level currently. Severe accident occur beyond the scope of design. Reconstruction of mind skills power employees, such as vigilance is the top priority. It should regain the confidence of the affected areas in efforts to put the body of TEPCO to the affected areas. Municipality and province is losing the trust of residents. That TEPCO interact directly with residents is important. The more than 1000 trillion yen, debt of Japan is increasing 40 trillion yen debt every year. Primary alternative is a 10 trillion yen them. Japan should not be a debt of more. It should be explained effectively to restart opposite the government accounts for half of the people.
Hagiwara, Hiroyuki; Yamada, Yumi*; Eto, Masao*; Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
The self-actuated shutdown system (SASS), which is selected for Japan Sodium-cooled Fast Reactor (JSFR), is a passive reactor shutdown system utilizing a Curie point electromagnet (CPEM). With CPEM, an excessive fuel outlet temperature rise is sensed and the control rods are released into the core, and the reactor can be shutdown. Therefore it is important for feasibility of SASS to be established by assuring a quick response of CPEM to the coolant temperature rise. In this paper, a device named "flow collector", which collects flows discharged from six fuel subassemblies surrounding CPEM backup control rods, has been proposed to ensure a shorter response time.
Hayakawa, Satoshi*; Ishikura, Shuichi*; Watanabe, Osamu*; Kaneko, Tetsuya*; Yamano, Hidemasa; Tanaka, Masaaki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
The present methodology was applied to the analysis for the 1/3-scale experiment of the hot-leg pipe of JSFR, and the predicted stress values were compared with the measured stress values. The predicted stress values were underestimated in the case of using the intact pressure fluctuations obtained by the unsteady fluid flow analysis. Therefore, the improvement of the prediction accuracy of the pressure fluctuations on the pipe wall was attempted.
Hamada, Noriaki*; Shiina, Koji*; Fujimata, Kazuhiro*; Hayakawa, Satoshi*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In a sodium-cooled fast reactor, a vortex cavitation evaluation methodology was developed to predict a possible cavitation generated by vortex at the center of accelerating flow. This methodology was applied to a scaled model experiment, leading to the prospect that the cavitation can be predicted.
Yamano, Hidemasa; Tanaka, Masaaki; Kimura, Nobuyuki; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Nuclear Engineering and Design, 241(11), p.4464 - 4475, 2011/11
Times Cited Count：19 Percentile：80.18(Nuclear Science & Technology)This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in JSFR, in particular emphasizing on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. Experimental efforts have been made using 1/3-scale and 1/10-sca1e single elbow test sections for the hot-leg piping and 1/4-scale and 1/7-scale triple-elbow test sections for the cold-leg piping. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced the pressure fluctuations onto the pipe. Simulation activities include Unsteady Reynolds Averaged Navier Stokes equation (U-RANS) and Large Eddy Simulation (LES) approaches. Numerical results using the U-RANS approach appear in this paper, indicating its applicability to the hot-leg piping experiments.
Yamano, Hidemasa; Tanaka, Masaaki; Kimura, Nobuyuki; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 18 Pages, 2009/09
This paper describes the current status of flow-induced vibration evaluation methodology development for the primary cooling pipings in JSFR, in particular emphasizing on R&D activities that investigate unsteady hydraulic characteristics in a short-elbow piping. Experimental efforts have been made using 1/3-scale and 1/10-scale single-elbow test sections for the hot-leg piping and 1/4-scale and 1/7-scale triple-elbow test sections for the cold-leg piping. Simulation activities include Unsteady Reynolds Averaged Naviar Stokes equation (U-RANS) approach with Reynolds Stress Model (RSM) using a CFD code and Large Eddy Simulation (LES) approach using in-house codes. Numerical results appears in this paper, focusing on its applicability to the hot-leg piping experiments. The numerical results could be provided to the input data for the structural vibration evaluation of the piping. The procedure of the flow-induced vibration evaluation is also described in this paper.
Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9296_1 - 9296_11, 2009/05
In the present study, three-dimensional thermal-hydraulic analyses of the reactor upper plenum in JSFR were applied to evaluating the following countermeasures including the geometrical structure. The basic ideas of countermeasures are as follows. (1) In order to mitigate the thermal striping phenomenon, locations where hot sodium discharged from fuel assemblies meets with cold sodium from control rods and/or radial blanket subassemblies are kept away from un-replaceable structures above the core. (2) In order to prevent the vortex cavitations, asymmetric flow in the upper plenum due to the radial slit with UIS is mitigated by installing a cylindrical structure named as dummy plug instead of the fuel handling machine only used for refueling period. Furthermore, flow holes on perforated plates with UIS are extended to mitigate radial sodium jet from lower part of UIS. (3) In order to prevent the cover-gas entrainment, Dipped Plate (DP) is installed below the sodium free surface. Original design of DP was a double wall type with many labyrinth seals, so the manufacturing regarded as difficult. Then a single wall type DP has been newly designed and examined in this analyses.
Aizawa, Kosuke; Nakanishi, Shigeyuki; Yamano, Hidemasa; Kotake, Shoji; Hayakawa, Satoshi*; Watanabe, Osamu*; Fujimata, Kazuhiro*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
To evaluate the flow-induced vibration in the actual-sized pipings of JSFR, computer simulation is necessary. In this study, as the first step, sensitivity analysis of turbulence flow models for unsteady short-elbow pipe flow has been carried out with the STAR-CD thermal-hydraulic simulation code. Through the sensibility analysis, the objective of this study is to propose the best analysis models which can reproduce the unsteady characteristics obtained in the 1/3-scale test results with 9.2 m/s of main flow. In this study, to take into account anisotropic characteristics of turbulence, two turbulent flow models were used: large eddy simulation (LES) and Reynolds stress model (RSM). The both validated simulations have reproduced flow separation region and periodic vortex shedding. The simulation results with both models were compared with power spectrum densities of pressure fluctuations which were used in the pipe vibration evaluation. Only the RSM simulation with the best combination has reproduced the pressure-fluctuation power spectrum densities, which were characterized by a peak frequency of 10 Hz in the 1/3 test with 9.2 m/s.
Hatae, Takaki; Naito, Osamu; Kitamura, Shigeru; Sakuma, Takeshi*; Hamano, Takashi*; Nakatsuka, Masahiro*; Yoshida, Hidetsugu*
Journal of the Korean Physical Society, 49, p.S160 - S164, 2006/12
no abstracts in English
Minato, Kazuo; Hayashi, Hirokazu; Mizuguchi, Koji*; Sato, Takeyuki*; Amano, Osamu*; Miyamoto, Satoshi*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.778 - 781, 2003/11
The simulation technology for the pyrochemical reprocessing of oxide fuels was developed to analyze experimental data, to predict experimental results, and to propose adequate conditions and processes. The simulation method was based on calculations of chemical equilibrium and electrochemical reactions. Some model calculations to simulate the experimental results were made on the process of electro-codeposition of UO
and PuO. Although it was difficult to trace the experiments and compare the calculated results with the experimental results quantitatively due to the limitation of available data on the experimental conditions, the calculated results were consistent with the experimental results. The phenomena of the repeated oxidation-reduction reactions between Pu
and Pu
ions and those between Fe and Fe
ions were theoretically analyzed,which caused the low current efficiency in the electro-codeposition process.
Shibata, Keiichi; Kawano, Toshihiko*; Nakagawa, Tsuneo; Iwamoto, Osamu; Katakura, Junichi; Fukahori, Tokio; Chiba, Satoshi; Hasegawa, Akira; Murata, Toru*; Matsunobu, Hiroyuki*; et al.
Journal of Nuclear Science and Technology, 39(11), p.1125 - 1136, 2002/11
Times Cited Count：669 Percentile：96.97(Nuclear Science & Technology)Evaluation for JENDL-3.3 has been performed by considering the accumulated feedback information and various benchmark tests of the previous library JENDL-3.2. The major problems of the JENDL-3.2 data were solved by the new library: overestimation of criticality values for thermal fission reactors was improved by the modifications of fission cross sections and fission neutron spectra for 
U; incorrect energy distributions of secondary neutrons from important heavy nuclides were replaced with statistical model calculations; the inconsistency between elemental and isotopic evaluations was removed for medium-heavy nuclides. Moreover, covariance data were provided for 20 nuclides. The reliability of JENDL-3.3 was investigated by the benchmark analyses on reactor and shielding performances. The results of the analyses indicate that JENDL-3.3 predicts various reactor and shielding characteristics better than JENDL-3.2.
Hatae, Takaki; Nakatsuka, Masahiro*; Yoshida, Hidetsugu*; Naito, Osamu; Kitamura, Shigeru; Sakuma, Takeshi*; Hamano, Takashi*; Tsukahara, Yoshimitsu
Proceedings of 29th European Physical Society Conference on Plasma Physics and Controlled Fusion, 4 Pages, 2002/00
no abstracts in English
Sato, Nobuaki*; Takenaka, Toshihide*; Hayashi, Hirokazu; Amano, Osamu*; Kawamura, Fumio*
no journal, ,
no abstracts in English
Niwa, Hajime; Kurisaka, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Kamiyama, Kenji; Yamano, Hidemasa; Miyahara, Shinya; Ohno, Shuji; Seino, Hiroshi; Ishikawa, Hiroyasu; et al.
no journal, ,
In order to develop the core damage evaluation technology (level 2 PSA) for sodium-cooled fast reactors, we develop the new analysis codes of post accident material relocation phase and of ex-vessel events, and we develop the technical bases that is necessary for level 2 PSA. In this presentation, summary and scope of the entire study is introduced as a part of the 4 series presentations.
Yamano, Hidemasa; Kamide, Hideki; Ohshima, Hiroyuki; Uto, Nariaki; Kotake, Shoji; Watanabe, Osamu*
no journal, ,
This paper reports the progress and plan of experimental and analytical studies on flow-induced-vibration characteristics in a large-diameter elbow pipe in order to ensure the validity of a sodium-cooled fast reactor design that aims at large-scale reactor power under a two-loop configuration in the cooling system.
Nakai, Ryodai; Kurisaka, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Kamiyama, Kenji; Yamano, Hidemasa; Miyahara, Shinya; Ohno, Shuji; Seino, Hiroshi; Ishikawa, Hiroyasu; et al.
no journal, ,
To develop a core damage evaluation technology (level-2 PSA) in sodium-cooled fast reactors, a new analysis method is developed for core-material relocation phase and internal containment vessel event. This study also develop technical basis necessary for the level-2 PSA.
Aizawa, Kosuke; Yamano, Hidemasa; Uto, Nariaki; Kotake, Shoji; Watanabe, Osamu*; Fujimata, Kazuhiro*
no journal, ,
A conceptual design study of a large-scale sodium-cooled fast reactor adopts a two-loop primary cooling system with large-diameter piping in order to reduce plant construction cost. In this design, one of issues is a flow-induced vibration behavior of the piping under a high Reynolds number of 10
order. To evaluate the piping integrity, it is necessary to obtain power spectrum densities of pressure fluctuations on the piping wall by a numerical unsteady flow simulation. In this study, the numerical simulation capability of Reynolds stress model and large-eddy simulation in the STAR-CD code has been investigated using the 1/3-scale hot-leg test data. Through the sensitivity analysis, the Reynolds stress model with appropriate analytical models has shown the best applicability to flow dynamics simulation in the short-elbow pipe.
Sakuma, Takeshi; Hatae, Takaki; Kitamura, Shigeru; Hamano, Takashi; Naito, Osamu
no journal, ,
no abstracts in English
